ML25143A051
| ML25143A051 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 06/10/2025 |
| From: | Byrd T NRC/NRR/DORL/LPL4 |
| To: | Boyce M Wolf Creek |
| Lee S, 301-415-3168 | |
| References | |
| EPID L 2024-LLA-0125, EPID L 2024-LLE-0026 GSI-191 | |
| Download: ML25143A051 (1) | |
Text
June 10, 2025 Mr. Cleveland Reasoner Chief Executive Officer and Chief Nuclear Officer Wolf Creek Nuclear Operating Company P.O. Box 411 Burlington, KS 66839
SUBJECT:
WOLF CREEK GENERATING STATION, UNIT 1 - AUDIT
SUMMARY
FOR LICENSE AMENDMENT REQUEST AND EXEMPTION REQUEST FOR A RISK-INFORMED RESOLUTION TO GENERIC SAFETY ISSUE-191 (EPID L-2024-LLA-0125 AND EPID L-2024-LLE-0026)
Dear Mr. Reasoner:
By letter dated September 12, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24260A071), Wolf Creek Nuclear Operating Company (WCNOC, the licensee) submitted a license amendment request, exemption request, and updated response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, dated September 13, 2004 (ML042360586), for the Wolf Creek Generating Station, Unit 1 (Wolf Creek). The amendment would modify the Wolf Creek licensing bases, including the affected portions of the technical specifications and updated final safety analysis report. Specifically, the amendment would allow the use of a risk-informed approach to address safety issues discussed in U.S. Nuclear Regulatory Commission (NRC) Generic Safety Issue 191, Assessment of Debris Accumulation on PWR [Pressurized-Water Reactor] Sump Performance.
To support its review, the NRC staff conducted a virtual regulatory audit on April 30 and May 8, 2025. The NRC staff reviewed documents and held discussions with members of WCNOC and its contractors. The regulatory audit summary is enclosed with this letter.
If you have any questions, please contact Thomas Byrd at (301) 415-3719 or by email at Thomas.Byrd@nrc.gov.
Sincerely,
/RA/
Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482
Enclosure:
Audit Summary cc: Listserv
Enclosure OFFICE OF NUCLEAR REACTOR REGULATION REGULATORY AUDIT
SUMMARY
IN SUPPORT OF LICENSE AMENDMENT REQUEST FOR RISK-INFORMED RESOLUTION TO GENERIC SAFETY ISSUE-191 WOLF CREEK NUCLEAR OPERATING COMPANY WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482
1.0 BACKGROUND
By letter dated September 12, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24260A071), Wolf Creek Nuclear Operating Company (WCNOC, the licensee) submitted a license amendment request (LAR), exemption request, and updated response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, dated September 13, 2004 (ML042360586), for the Wolf Creek Generating Station, Unit 1 (Wolf Creek). The amendment would modify the Wolf Creek licensing bases, including the affected portions of the technical specifications (TSs) and updated final safety analysis report.
Specifically, the amendment would allow the use of a risk-informed approach to address safety issues discussed in U.S. Nuclear Regulatory Commission (NRC) Generic Safety Issue-191, Assessment of Debris Accumulation on PWR [Pressurized-Water Reactor] Sump Performance.
On January 14, 2025 (ML25010A369), the NRC staff issued an audit plan, which provided the list of requested documents and other details pertaining to the audit. An audit team, consisting of NRC staff and two contractors from the Southwest Research Institute (SwRI), conducted a remote regulatory audit to support the review of the requested licensing actions on April 30 and May 8, 2025. The purpose of the audit was to gain an understanding of the information needed to support the NRC staffs licensing decision regarding the licensing actions and to identify additional information required on the docket by the NRC staff to make a regulatory finding. Any additional information provided to support the licensing review will be evaluated by the NRC staff and included in its final findings.
The regulatory audit is a planned license or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. The regulatory audit is conducted with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. Performing a regulatory audit of the licensees information is expected to assist the NRC staff in efficiently conducting its review or gain insights into the licensees processes or procedures.
Information that the NRC staff relies upon to make the safety determination must be submitted on the docket. However, the NRC staff may review supporting information retained as records under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.71, Maintenance of records, making of reports, and 10 CFR 54.37, Additional records and recordkeeping
requirements, which, although not required to be submitted as part of the licensing action, would help the NRC staff better understand the licensees submitted information.
2.0 AUDIT ACTIVITIES The NRC audit team consisted of the NRC staff members and the contractors from SwRI listed in section 5.0 of the audit plan (ML25010A369). The NRC staff developed questions based on its review of the licensees submittal (ML24260A071) and the information provided in response to the initial audit items list in the audit plan. The licensee provided responses to the NRC questions on the electronic portal. These responses provided the starting point for discussions during the audit. The NRC audit team held an initial remote meeting to discuss the issues identified in the audit plan on April 30, 2025, with the licensees staff and contractors. The NRC staff and its contractors held a follow-up meeting with the licensee on May 8, 2025. The list of WCNOC participants is provided in attachment 1. During the audit, the NRC audit team participated in technical discussions with the licensee to meet the objectives listed in section 3.0 of the audit plan.
3.0 RESULTS OF THE AUDIT Technical discussions were focused on the following major areas identified by the technical review branch responsible for the specific areas of review. Attachment 2 provides the summary of the technical discussions.
APLB, Probabilistic Risk Assessment (PRA) Licensing Branch B (risk analysis and methodology)
NVIB, Vessels and Internals Branch (piping and inspections)
STSB, Technical Specifications Branch (TS changes and debris methodology/effects)
NCSG, Corrosion and Steam Generator Branch (coatings and chemical effects)
ESEB, Structural, Civil, Geotechnical Engineering Branch (strainer structural)
APLC, PRA Licensing Branch C (seismic analysis)
The NRC audit team participated in audit meetings with the licensee on April 30 and May 8, 2025. During these meetings, the NRC and licensee discussed areas identified by the NRC as detailed in attachment 2 of this summary.
At the end of the meetings, the NRC staff provided a brief conclusion based on the discussions.
There were no open items in the discussion and no deviation from the audit plan. The licensee agreed to provide a supplement to the application to address audit discussion points.
The information submitted in support of the Wolf Creek LAR is under review. If any additional information needed to support the LAR review is required after the supplemental information is received, it will be formally requested by the NRC staff in accordance with NRR Office Instruction LIC-115, Revision 1, Processing Requests for Additional Information, dated August 9, 2021 (ML21141A238).
Attachments
- 1. List of WCNOC Participants
- 2. Appendix - Summary of Discussions
LIST OF WCNOC PARTICIPANTS Wolf Creek Nuclear Operating Company, Wolf Creek Generating Station, Unit 1 Team John Harris, Principal Engineer and Generic Safety Issue (GSI)-191 Project Lead, Design Engineering David Alford, Principal Engineer, Nuclear Engineering Probabilistic Risk Assessment (PRA)
David Vu, Lead Engineer, Nuclear Engineering PRA Cody Haddock, Lead Engineering Technologist, Engineering Programs Katie Gibbon, Senior Engineer, Strategic Engineering Frank Galati, Senior Manager, Nuclear Engineering Josh Turner, Manager Nuclear Regulatory Affairs, Licensing Nathan Lee, Lead Engineer, Licensing Enercon - GSI-191 Contractor Tim Sande, PRA Engineering Manager Kyle Holstein, Lead Risk Applications Engineer Firat Alemdar, Chief Civil/Structural Engineering Manager Duygu Saydam, Chief Civil/Structural Engineering Supervisor
Appendix - Summary of Discussions In addition to the items discussed below, the U.S. Nuclear Regulatory Commission (NRC) staff reviewed information provided during the audit (Audit Plan dated January 14, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25010A369))
regarding a license amendment request (LAR), exemption request, and updated response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, dated September 13, 2004 (ML042360586), for the Wolf Creek Generating Station, Unit 1 (Wolf Creek). These reviews contributed to the NRC staff understanding of the issues discussed below. In addition, the NRC staff review of the data provided for the debris generation and transport for each break evaluated indicates that the model used for these parts of the analysis behaves as expected.
The listing below retains the original issues that were identified in the attachment to the audit plan and provides the NRC staff understanding of how the issues will be resolved, if necessary.
The NRC staff expects that the items that require a revision to requested licensing actions will be addressed via one or more supplements. The NRC will request additional information, as necessary, to attain the knowledge required to make its safety and regulatory conclusions.
The NRC staff did not reach any regulatory conclusions during the audit. Regulatory decisions will be based on information that has already been received on the docket or a future licensee supplement to the requested licensing actions.
Note that the page numbers referenced in this appendix are based on the PDF page numbers from the licensees submittal dated September 12, 2024 (ML24260A071). The questions are split up by technical branch areas.
Questions Discussed Probabilistic Risk Assessment (PRA) Licensing Branch B (APLB) Questions APLB-Q1 (Related to Audit Plan Initial Item No. 3)
The application uses the 25-year loss-of-coolant accident (LOCA) frequencies from NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the Elicitation Process, April 2008, rather than the 40-year values also provided in that document. The provided justification is that the Wolf Creek plant is not yet 40 years old and so the 25-year values are appropriate. However, the 40-year anniversary of the Wolf Creek Operating License will occur prior to the issuance of this license amendment, if the NRC staff determines that it is appropriate to issue it.
Discuss the effect that using the 40-year values would have on the results of the analysis, including on sensitivity studies included in the application.
Discuss whether this frequency change would move the proposed change over the threshold into Region II of Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions of Plant-Specific Changes to the Licensing Basis, figures of merit.
Discuss any plans to update the analysis once the 40-year anniversary is passed.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable but needs the response on the docket to finalize its regulatory finding.
APLB-Q2 Identify whether the risk-informed analysis will be included in the PRA maintenance program.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
Vessels and Internals Branch (NVIB) Questions NVIB-Q1 (Related to Audit Plan Initial Item No. 2)
Section 2.3.2, RCS [Reactor Coolant System] Pressure Boundary of attachment IX to ET 24-000478 (PDF page 304 of the LAR) states that Wolf Creek developed a program plan to manage the risk of Primary Water Stress Corrosion Cracking (PWSCC) degradation in Alloy 600 components and Alloy 82/182 welds. The plan is in accordance with 10 CFR 50.55a [Title 10 of the Code of Federal Regulations Section 50.55a], ASME Code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code]
Casesl N-722-1 ([Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials]) and N-770-2
([Alternative Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler material With or Without Application of Listed Mitigation Activities]),
and NEI [Nuclear Energy Institute] 03-08 (Guideline for the Management of Material Issues, Revision 3). The plan identifies all Alloy 600/82/182 locations, ranks the locations based on their risks of developing PWSCC, provides inspection requirements, and presents mitigation/replacement options. Wolf Creek has either implemented or planned mitigation measures for the welds of concern. Periodic inspections of the Alloy 600 components and Alloy 82/182 welds are covered in the ISI [inservice inspection] Program...
The licensee has uploaded the PWSCC degradation management plan in the ePortal. The NRC staff has reviewed the plan. The plan contains valuable information that the staff is seeking as shown in the audit question below. The staff needs the information as requested in the audit questions NVIB-Q2, Q3, and Q4 below to address the PWSCC degradation management issue.
Two options.
Option 1 - the licensee submits to the NRC its PWSCC management plan (the document) on the docket. In this scenario, the NRC staff would not ask audit questions NVIB-Q2, Q3, and Q4 regarding PWSCC in the request for additional information (RAI) questions because the staff could use the plan on the docket to address the PWSCC issue in the safety evaluation (SE).
Option 2 -The licensee does not submit the PWSCC management plan (the document) on the docket. In this scenario, the NRC staff will officially issue the PWSCC-related RAI questions as shown in the audit questions. The staff will use the licensees response to the RAI questions to evaluate the PWSCC issue in the SE.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding for Question NVIB-Q2.
NVIB-Q2 (Related to Audit Plan Initial Item No. 2)
(a) Provide/upload a list (e.g., a table) of and/or documents related to in-scope welds with associated piping system that are fabricated with nickel-based Alloy 82/182 material.
(b) On the list, identify in-scope Alloy 82/182 welds that have and have not been mitigated to reduce their susceptibility to stress corrosion cracking.
(c) Discuss whether the unmitigated in-scope Alloy 82/182 welds will be mitigated in the future.
(d) For those unmitigated in-scope Alloy 82/182 welds, discuss how they are being inspected (e.g., inspection frequency and method) to monitor ensure their structural integrity. (f)
Upload in the portal the NRC-approved relief requests regarding actions on Alloy 82/182 welds, if any.
The NRC staff reviewed the documents provided in response to Audit Plan Initial Item No, 2, The NRC staff requires additional information regarding the table notation in the portal response. This information should be provided on the docket along with the information provided in response to Audit Plan Initial Item 2.
NVIB-Q3 (Related to Audit Plan Initial Item No. 2)
(a) Provide a list of in-scope welds or components that contain indications/flaws that remain in service and that have not been repaired.
(b) Identify the in-scope welds that have been repaired and their condition after the repair.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
NVIB-Q4 (Related to Audit Plan Initial Item No. 2)
(a) If the risk analysis contains welds that contain flaws that have not been repaired, discuss whether the LOCA frequency for the degraded welds in the risk analysis is increased from that of NUREG-1829 estimates. If the LOCA frequency is not increased in the analysis, provide justification.
(b) Discuss whether the failure probability value in the risk analysis is increased to account for the in-scope unmitigated nickel-based Alloy 600/82/182 components (welds and pipe components). If a higher probability of failure was not used in the risk analysis, provide justification.
The NRC staff reviewed the documents provided on the portal. In addition to the portal response, the NRC staff requires additional information regarding the break sizes (list all breaks smaller and larger than the 10-inch threshold break size) that were evaluated in the risk-informed analysis. This information should be provided on the docket.
NVIB-Q5 The NRC staff notes that as part of plant operating procedures, pressurized water reactor owners inspect RCS piping and associated components in addition to the inspections performed per NRC regulations or the ASME Code,Section XI, such as operator walkdowns, opportunistic inspections, the boric acid corrosion program, and the fatigue monitoring program per Materials Reliability Program (MRP)-146, Revision 1, Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines. Discuss any administrative or opportunistic inspections at the Wolf Creek plant that monitor the structural integrity of the in-scope piping and components in addition to the NRC regulations.
The licensee provided a response to this question on the portal. The NRC staff reviewed the information and concluded that additional information regarding the boric acid control program, leak inspection program, and other opportunistic inspections regarding potential reactor coolant pressure boundary leakage be included on the docket.
NVIB-Q6 Nuclear plants have a leakage detection system to monitor leakage from the RCS. The RCS leakage detection system is usually designed in accordance with RG 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, to provide time for appropriate operator action to identify and address RCS leakage. However, based on operating experience, the RCS leakage detection system in some nuclear plants does not provide the same capability as originally designed. Discuss (a) any changes made to the current RCS leakage detection system that are different from the description in the Final Safety Analysis Report, (b) whether the capability of the sensors and instrumentation of the current RCS leakage detection system is consistent with RG 1.45, Revision 1, and (c) whether the diversity, redundancy, and reliability of the current detection instrumentations still satisfy the guidance in RG 1.45, Revision 1.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.
Technical Specification Branch (STSB) Questions STSB and the NRC contractors used the response to Audit Item No. 1 to verify that the licensees calculations for debris generation and transport were performed correctly.
STSB-Q1 For the adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-567, Add Containment Sump TS to Address GSI-191 Issues, it appears that TS 5.5.15, Safety Function Determination Program, should be revised per the traveler. This was not included in the LAR or
addressed as a variation to the traveler. Provide a supplement to include the change or explain why the change is not needed.
The licensee stated that it would provide a supplement to the LAR that will incorporate the change into its TS. The NRC staff concluded that this would provide an acceptable resolution to this issue.
STSB-Q2 Table B 3.6.8-1 in attachment V of the LAR (PDF page 59) lists the containment sump debris limits for breaks that are considered to pass the deterministic criteria. The table values are also explained starting on PDF page 315. Discuss how these values were developed. The bases for the fiber values were not clear and did not appear consistent with the other values. This may also be related to another question regarding the information on PDF page 316 of the LAR.
- a. The title implies that these are containment sump debris limits. Would strainer limits be a more accurate description based on the notes for the table?
- b. Should the Thermolag particulate value in the table be increased to bound or equal the transported value with a subsequent reduction in coatings margin. This might be a more realistic representation in design basis amounts and eliminate a footnote. The value in the Bases is slightly exceeded by the current plant condition as discussed on page 316, footnote 4. Alternately, should the TS table provide an explanation on this?
- c. The fiber fines value in the bases table appears to exceed the full debris load (FDL) tested amount by about 3 pounds (141.78 tested and 144.6 in the table). The fine fiber limit value in the Bases table is compared to a quantity for two train transport (in-vessel analysis?) on PDF page 315. To reflect the lowest fine fiber amount that can be accommodated by the strainer, should the strainer head loss test debris amount be used?
- d. Provide an explanation of what debris sizes and amounts the total fiber value includes and how it will be used in an operability analysis. It appears to be the same total fiber value that was included in the FDL test. The footnote for the Bases table states that it is the maximum amount allowed to transport to a single strainer. The footnote on PDF page 315 states that this is the amount included in the full debris load test. Is erosion considered in this value? How, during an operability assessment, would it be determined whether discovered fiber exceeds this or the fine fiber limit? If it is determined that the fine fiber limit, including erosion, is not exceeded and all additional fiber plus the fine fiber are less than the total fiber limit that operability would be demonstrated so this value is likely valid. It is just difficult to determine how it would be used.
- e. How was the value for degraded paint chips determined and how is the condition assessed if additional chips are identified in the plant?
- f.
The footnotes for the Bases table provide some insight into the above, but it is not clear how someone referring to the table would be able to determine whether a plant condition is bounded by the limits or easily determine whether this is the case or find the basis for the values.
The licensee provided a response to this question on the portal and the NRC staff reviewed the information. The information provided adequate clarification for some of the issues identified in this question, but the NRC identified issues that required additional discussion. During the audit discussions, the licensee provided an updated concept for determining operability and margin values for the various types of debris, particularly fibrous debris. The NRC requires this information to be provided as a supplement to the LAR. This response also covers questions related to STSB-Q14b and Q16.
STSB-Q3 In attachment VII of the LAR (PDF page 96), it is stated that control rod drive mechanism (CRDM) ejections and bottom head instrument penetration breaks are excluded as debris generation events due to the location of the breaks and the direction of any jet from these locations. The discussions in sections 3.a and 3.b of attachment VIII of the LAR (PDF pages 132-143 do not elaborate on why the CRDM housings and instrument penetrations are excluded or describe the insulation materials in the vicinity of the potential breaks. Discuss the basis for the elimination of these breaks. For these break locations, the NRC is primarily concerned with problematic insulation types that are not bounded by the current debris generation evaluation The licensee provided clarifying information on the portal. The NRC concluded that the break analysis is acceptable. However, the NRC requested that the licensee provide the type of insulation on the reactor vessel head on the docket.
STSB-Q4 In attachment VII of the LAR (PDF page 114), the parameter for containment spray switchover time states that the longest switchover time was used for the in-vessel effects analysis. The NRC staff understands that containment spray recirculation is assumed to fail for the in-vessel analysis beyond design basis case and that one pump fails for the design basis case. Both cases met the total in-vessel limit (PDF page 261). Please expand upon the discussion on PDF page 114 of the LAR. For example, is this assumption intended to have only the emergency core cooling system pumps injecting for the longest time, increase the pool volume (decrease debris concentration), etc.? What is the basis for choosing the maximum switchover time as limiting?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q5 On PDF page 149 a description of the transport logic tree calculations is provided. How were these calculations carried out? How was BADGER data inputted into the calculations? Was any interim processing required?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q6 In attachment VIII of the LAR (PDF page 165), the licensee states that for the two-train case erosion was applied to the small and large pieces of fiber that transport to the strainer. What is the reason for applying the erosion factor to the transported fiber in this case and not the single train case?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q7 On PDF page 172 of the LAR, explain how the transport fractions for small and large NUKON eroded fine particles were developed. Do these values represent the number of fine particles created by erosion transported, the amount of small and large debris transported as fine particles, or something else?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q8 In attachment VIII of the LAR (PDF page 179), the vortex testing is described. Provide the water level above the top the strainer that was maintained during the testing. Is this the same value provided on page 187 for the submergence level? (14 to 18 inches) The value for submergence appears to be about 7 inches. How does this compare to the submergence during testing? How does this value relate to submergence of the strainer top surface? Does any difference between the plant and test submergence of the plenum need to be considered? Discuss whether the submergence level in the plant increases due to Core Spray injection after the point at which the minimum water level is calculated and what the magnitude of this increase would be.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q9 In attachment VIII of the LAR (PDF page 185), chips representing degraded qualified coatings are discussed. The NRC staff could not identify a discussion of degraded qualified coatings in the coatings section (3.h) on PDF page 221. Describe the purpose of the coating chips used in the testing. Paint chips are listed in the Sump Debris Limit, Table 1 of Attachment X, PDF page 315, (maximum transportable to one strainer), and in Attachment V, PDF page 59. The NRC staff calculated that the thickness of the coating would be 8.33 millimeters if it equaled 0.11 cubic feet (ft3) as shown in table 3.f.7-1 on PDF page 197 of the LAR. If degraded paint chips are discovered in containment what transport metrics will be used? How was the 0.11 ft3 value determined? Is the degraded qualified coatings value included in testing and the tables discussed above used only for margin?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q10 The test values in table 3.f.4-2 on PDF page 188 of the LAR appear to be higher than the transported values reported on PDF page 174. What was considered when choosing these test amounts?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q11 The degasification analysis, which begins on PDF page 207 of the LAR, states that the void fraction value is determined at the midpoint of the strainer. This is acceptable if voids occur across the full height of the strainer. If the void fraction is zero at the bottom of the strainer due to added submergence using the midpoint will result in a non-conservative value. Discuss the acceptability of the use of the strainer midpoint for the Wolf Creek degasification analysis.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q12 In attachment VIII of the LAR (PDF page 234, the weight of debris accumulated on the strainer is discussed as an input for the structural calculation. Provide the weight of debris assumed in the analysis.
The licensee provided a response to this question on the portal that included the debris amount that collects on each strainer module. The NRC staff concluded that the response was acceptable and requested that the information be provided on the docket.
STSB-Q13 In attachment VIII of the LAR (PDF page 236), the NRC was unsure how to interpret the last sentence of the second bullet in section 3.k.2. Clarify whether the structural analysis uses the greater low temperature differential pressure (dP) of 5.5 feet for all cases or if the reduced dP of 4.0 feet is used for some cases. Both values are listed on PDF page 235. Was the thermal stress at 268 degrees Fahrenheit used for all cases?
The licensee provided a response to this question on the portal. The NRC staff reviewed the question and requested additional clarification during the audit discussions. The clarifications provided the information necessary for the NRC to understand how the calculations were performed. The NRC staff requested that the material properties used for the low and high temperature structural cases be provided on the docket. This information could be added to the second bullet on PDF page 236 of the LAR (section 3.k.2)
STSB-Q14
- a. In attachment VIII of the LAR (PDF page 257), the licensee states that only the low-fiber penetration curve was used as shown later in the section. Discuss this statement. Was this assumed because the low fiber curve bounded the debris load that could occur from the threshold break size (or other less conservative analysis assumptions)?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
- b. For the evaluation of fiber inside the reactor vessel on PDF page 260, what is the fiber amount assumed in the pool at the beginning of the calculation, including latent fiber? Is all fine fiber assumed. If small or large fiber is assumed, how is erosion treated?
The licensee provided a response to this question on the portal. The NRC staff discussed the response with the licensee. This issue is related to questions STSB-Q2. The licensee will provide information on the docket that will address this issue along with the response to STSB-Q2.
STSB-Q15 In attachment X of the LAR (PDF page 314), the licensee states that the new design basis for Wolf Creek will be keeping the risk values in Region III of RG 1.174. Should this be reflected in attachment VII, section 6.0, Reporting and Corrective Actions (PDF page 122), or is this covered conservatively by the debris limits? Other plant changes might result in exceeding the RG thresholds without an increase in debris amounts.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
STSB-Q16 The NRC could not verify all the values in the debris margins table (table 2 in attachment X (PDF page 316). The NRC also did not understand all the footnotes for the table.
- a. The value for total fiber fines, small pieces, and large pieces appears to include small and fine Nukon fiber and latent fiber. Does this value include ThermoLag fiber? The footnote (2) for this value also states that the built-in margin for latent fiber is subtracted.
It is not clear what this means.
- b. Should the ThermoLag particulate value in the table be increased to bound or equal the transported value with a subsequent reduction in coatings margin. This might be a more realistic representation in design basis amounts and eliminate a footnote. The current value is slightly greater than the limit in footnote 4. It might be better to use some coatings margin to offset the slight negative margin in this value. (Question also asked related to the TS Bases table on PDF page 59.).
- c. The limit values for fine and total fiber were questioned with respect to the TS Bases table on PDF page 59.
The licensee provided a response to this question on the portal. The NRC staff discussed the response with the licensee. This issue is related to questions STSB-Q2. The licensee will provide information on the docket that will address this issue along with the response to STSB-Q2.
Corrosion and Steam Generator Branch (NCSG) Questions NCSG-Q1 Table B.3.6.8-1 (PDF page 59 of the LAR) identifies containment sump debris limits for breaks less than or equal to 10 inches. Aluminum is not included in this table. How is aluminum tracked to ensure assumed analysis amounts are not exceeded?
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.
NCSG-Q2 Please clarify the minimum containment sump pool pH. In attachment VIII of the LAR (PDF pages 263 and 274, the licensee states that the Wolf Creek minimum final containment pool pH is 8.78. Table 3.n.1-2 on PDF page 265 indicates the minimum sump pH (long-term) is 8.5. If the appropriate value is 8.5, discuss how that would affect the precipitation temperatures that were determined for the sump strainer and the reactor vessel.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.
NCSG-Q3 Do WCAP-17788-P tests 07-05 and IBOB 07-05 have test conditions that are representative for the Wolf Creek post-LOCA conditions? If so, what is the reason they are not referenced in the LAR?
The licensee provided a response to this issue on the portal. The NRC discussed this issue with the licensee during the audit call and requested clarification on the amounts of e-glass included in the tests. The licensee responded that the amounts of e-glass that would remain in the containment were not fully quantified at the time of the tests. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.
Structural, Civil and Geotechnical Branch (ESEB) Questions ESEB-Q1 In attachment VIII of the LAR (PDF page 235), there is discussion of required strength and yield strength. Because steel loses strength with increasing temperature with elevated temperatures, clarify if temperature effect on the yield strength is considered.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
PRA Licensing Branch C (APLC) Questions APLC-Q1 In attachment VII of the LAR (PDF page 91), the licensee states, Seismic events can result in direct or indirect LOCAs that generate and transport debris similar to a random pipe break LOCA. A direct seismically induced LOCA occurs when the RCS pressure boundary fails due to seismic force[s]. An indirect seismically induced LOCA occurs when a support or structure fails due to seismic forces, which subsequently causes an RCS pressure boundary failure.
However, in Section 2.3.4, Seismically Induced LOCAs, in attachment VII of the LAR (PDF page 98), it appears that the licensee only addresses indirect LOCAs caused by support failures, not by structure failures.
Provide examples of indirect LOCAs resulting from structural failures and clarify whether such failures were considered in the licensees Generic Safety Issue (GSI)-191 evaluation. If they were not considered, provide justification for their exclusion.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.
APLC-Q2 In attachment VII of the LAR (PDF page 92), the licensee states, High wind events, including tornados, would not generate debris inside containment and therefore are screened from the GSI-191 risk quantification. While containment can withstand the effects of high winds including tornado missiles, high winds may still lead to LOCAs through loss of offsite power (LOOP),
potentially resulting in debris blockage.
Discuss whether LOCAs due to LOOP caused by high winds should be considered in the licensee's GSI-191 risk evaluation.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.
APLC-Q3 In attachment VII of the LAR (PDF page 95), the licensee states:
The WCGS internal events PRA model uses the following definitions for small, medium, and large LOCAs:
Small LOCA = 0.5 to 2-inch breaks
Medium LOCA = 2 to 6-inch breaks Large LOCA = Greater than 6-inch break It appears that the LAR does not specify whether these LOCA size definitions apply to other hazards.
Clarify if the same LOCA size definitions used in the internal events PRA model are also applied to other hazards, including seismic events. If different definitions are used, provide justification for the variation.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
APLC-Q4 In attachment VII of the LAR (PDF page 101), Section 2.4.2, Risk Contribution from Relevant Hazards, the licensee discusses the quantified risk attributable to debris in terms of change in core damage frequency and large early release frequency or CDF and LERF. The LERF contribution is calculated using the conditional large early release probability (CLERP) for a large LOCA that results in core damage, with a CLERP value of 2.82E-05. However, it is noted that South Texas Project, Units 1 and 2 (STP) used a CLERP value of 2.5E-3 in its GSI-191 risk-informed evaluation (NRC staff SE, page 83, ML17038A223).
(a) Given that both STP and WCGS are Westinghouse 4-loop plants, their CLERP values would be expected to be similar. Explain the significant difference in the CLERP values used for the two plants.
(b) It appears that the same CLERP value is applied to both internal events and seismic events. However, due to differences in failure characteristics, seismic-induced LOCAs may require a different CLERP than those from internal events. Justify the use of the same CLERP for both internal and seismic events.
The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and provided the clarifying information necessary. No further information is required on the docket.
APLC-Q5 Audit document, WCN021-CALC-008, figure 1, presents the WCGS seismic hazard curves for various spectral frequencies, including the peak ground acceleration (PGA). The figure title suggests that the PGA is defined at 100 hertz (Hz), but the figure itself includes separate curves for both the PGA and 100 Hz. Clarify which frequency corresponds to the PGA.
The NRC staff discussed this issue with the licensee during the audit call and requested that the licensee clarify the spectral frequency associated with the PGA used in the GSI-191 resolution and state whether the seismic-induced LOCA frequencies are impacted by the clarification in spectral frequency. If LOCA frequencies are affected, the licensee will provide revised seismic-induced LOCA frequency estimates accordingly. The NRC also requested that the licensee provide the reference Electric Power Research Institute document on the portal to allow the review of this reference. This information is required on the docket.
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